In order to perform the various experiments with a bubble chamber, a high intensity flash tube is used. This report briefly describes the power supplies designed and constructed to power these lamps.

The power supplies designed and constructed to power high intensity flash tubes used in bubble chamber experiments are briefly described and are accompanied by a schematic diagram of the layout. (D.C.W.)

Space has been reserved near the southeast corner of the 100-N Area for the 122-N Decontamination Facility. Previous correspondence between Burns and Roe, Inc and General Electric bae discussed various facilities which might be needed in the building. The concepts of the decontamination processes are under active development by research groups at Hanford. At present, there are several workable processes known; each one has one or more fairly serious drawbacks.

Water treatment systems for preparing and maintaining high purity water in out-of-reactor or in-reactor test oops are becoming increasingly important. In out0of-reactor experiments the presence of ionic impurities in the water has a marked influence on film formation and corrosion rates. It is therefore , imperative that these impurities be maintained at the lower practical concentration.

The ANGIE, one of a series of reactor neutronic programs for an IBM 709 or 7090 data processing system, solves the time-dependent, multi-group, neutron diffusion equation for 1 to 26 energy groups applied to a rectangular mesh superimposed on either an x-y or an r-z plane.

The objective of the Accurate Nuclear Fuel Burnup Analyses program is to develop more accurate methods for burnup analysis for general use than the current method of analysis of Ca-137 or Sr-90. The program will require from three to five years of effort.

The rapid advances in many technologically important areas have not only served to accentuate the increased demands for high temperature metals and ceramics but have necessitated a more through knowledge of their physical properties when exposed to high temperature service. Toward this latter end, the use of X-ray diffraction has proved an invaluable tool in providing data of regions of thermal stability, expansion coefficients, solid solubility limits, and phase transformations by direct examination at temperature. Since this Laboratory has for some time now been engaged in the study of refractory nuclear materials, it was thought desirable to employ and possibly extend this technique to temperatures ranging up to 3000°C. This communication will describe the equipment developed for this purpose, with experimental results to be described in subsequent publications.

Technical report describing a UO2-Mo cermet fuel assembly fabricated for long-term irradiation performance testing in the Vallecitos Boiling water Reactor. The design and fabrication histories of this assembly are described and pre-irradiation data on each individual rod are presented. Molybdenum was added to improve the bulk thermal conductivity of the fuel, so that fuel temperatures would remain comparatively low during high-power level operation of the fuel element. The molybdenum was incorporated into the compacts either as fibers or as a thin coating on individual UO2 particles. Fuel pellets were produced from these materials by vacuum hot pressing. The distribution of the molybdenum in both types of cermet fuels appeared favorable to good heat transfer. The fibers were oriented predominantly in the radial planes of the pellet as a result of the uni-directional compaction during the hot-pressing operation. In the pellets made from the coated particles, a continuous network of molybdenum occurred as a result of the coating welding together during the hot-pressing operation. The test assembly contains eight fuel rods; three contain UO2-Mo cermet, three contain the cermet produced from the coated particles, and two are for reference and contain the conventional sintered UO2 pellet fuel. The nominal outside diameter …

This quarterly progress report details the ongoing research and experiments at the Oak Ridge National Laboratory as part of the Aircraft Nuclear Propulsion Project. The first part of this report discusses reactor theory and design. The second part of this report is not included. The third part of this report discusses materials research. The fourth part of this report includes appendixes

This quarterly progress report of the Aircraft Nuclear Propulsion Project at ORNL records the technical progress of the research on circulating-fuel reactors and other ANP research at the Laboratory under its Contract W-7405-eng-26. The report is divided into three major parts: I. Reactor Theory, Component Development, and Construction, II. Materials Research, and III. Shielding Research. The ANP Project is comprised of about 530 technical and scientific personnel engaged in many phases of research directed forward the achievement of nuclear propulsion of aircraft. A considerable portion of this research is performed in support of the work of other organizations participating in the national ANP effort. However, the bulk of the ANP research at ORNL is directed toward the development of a circulating-fuel type of reactor. The design, construction, and operation of the Aircraft Reactor Test (ART), with the cooperation of the Pratt & Whitney Aircraft Division, are the specific objectives of the project. The ART is to be a power plant system that will include a 60-Mv circulating-fuel reflector-moderated reactor and adequate means for heat disposal. Operation of the system will be for the purpose of determining the feasibility, and the problems associated with the design, construction, and operation, of a …

"The energy method is applied to study the stability of two types of difference approximations to parabolic partial differential equations, the alternating direction methods Douglas, Peaceman, and Rachford, and a new semi- explicit method. Each difference scheme is proved to be unconditionally stable. These results apply to parabolic equations with variable coefficients, defined in cylindrical domains with an essentially arbitrary bounded base."

The angle by which a magnet deflects the trajectory of a particle in the x-z plane is proportional to the integral [integral not transcribed] taken over the trajectory wherever B≠0. Alternating gradient focusing is achieved by designing magnets so that I=I(x) varies linearly with x over a suitable x interval. Usually this is done by shaping the poles to give a linear variation of By with x while keeping the length of the magnet constant for different x. Certain advantaged may be gained by varying the effective length of the magnet with x and keeping By constant so that the integral varies properly with x. Figure 1 shows several such poles for which the trajectory length, and hence the integral (1), varies approximately from 2/3 to 4/3 of the mean value.

From introduction: This report covers an investigation of solid-state bonding as a technique for aluminum cladding uranium plates of 3 by 0.180-in. cross section in lengths up to 14 ft.

As there is no accurate measurements of the energy distribution of neutrons emerging from large uranium blocks, it becomes of importance to see how much information of this kind can be derived from the slowing down distribution of the neutrons in a water tank.

From introduction: "The concluding remarks of the report on the electron diffraction study of UF6 suggest two alternate explanations for the apparent discrepancy between the structure proposed for this compound as deduced from X-ray diffraction work on single crystals and from the electron diffraction investigation of the vapor. The second explanation is based on the premise that the difference may be due to essentially unlike methods of treating the diffraction data. In this report we shall analyze X-ray powder photographs following a scheme parallel to that used in the analysis of the electron diffraction data, to check whether this particular explanation is tenable."

Technical report covering experiments happening on the Analytical Chemistry Division's sites at the Oak Ridge National Laboratory. Includes information on ionic analyses, radiochemical analyses, spectrochemical analyses, service analyses, inorganic preparations, analytical chemical control of homogeneous reactor solution, optical and electron microscopy, and service analyses. [From Abstract]

In an elastic collision the neutron loses part of its kinetic energy to the nucleus with both the kinetic and momentum of the system being conserved. However, for many elements the scattering is not isotropic in the center-of-mass system at the higher neutron energies. Many of the present reactor multigroup codes include anisotropic scattering at the high neutron energies, while many others assume isotropic scattering at all energies. In order to consider some of the effects of including anisotropic scatting, reference is made t the multigroup equation generally assumed for the slowing-down density.

Well-type scintillation counters have found use in many radiochemical laboratories. A gamma scintillation well counter has been used to measure the gamma activity of liquid samples in the 325 Building counting room for about a year. This well counter has been built and calibrated so that gamma activity measurements made with it can be easily compared with measurements made with the gamma scintillation counter (GSC). The well counter is electronically identical to the present GSC and differs only in the shape of the crystal used and the physical arrangement of the lead shield. The crystal contains a well which allows a tube containing the sample to be inserted in the crystal. The physical arrangement of the detector greatly simplifies the preparation of liquid samples for activity measurement. The 325 Building well counter and its application to chemical research and plant process analysis will be discussed in the following paragraphs. An operating procedure is also included.

Introduction: In anticipation of the design of equipment for heat transfer tests, a series of short exposure corrosion tests was initiated in an effort to determine the order of magnitude of the corrosion involved. Information was required on the effect on corrosion rates of applying a potential between two concentric tubes, the annulus between them containing condensate water.

A description is given of a method suitable for the automatic solution of certain optimum design problems on a digital computer for cases where the number of constraints imposed on the design is not greater than the number of design variables. The problem is transformed to one requiring the minimization or maximization of an unconstrained function, for which a gradient method is used.

The information concerning beryllium assembled herein has been selected primarily to provide information that might be pertinent to the design of the ART. The creep-rupture data in the literature have been obtained from vacuum cast and extruded beryllium, rather than from the hot-pressed beryllium which will be used in the ART. A research program is being conducted by the Brush Beryllium Company, under contract to the Oak Ridge National Laboratory, which will eventually provide the design engineers with data on the high temperature strength properties of hot-pressed beryllium. In the interim, it will be necessary for the metallurgists and design engineers to make judicious estimates and extrapolations from the data available.

"This is a bibliography of journal articles relating to beryllium oxide. Sources consulted include; Chemical Abstracts, 1951-1955; Physics Abstracts, 1951-1955; Industrial Arts Index, 1955-Jan, 1957; and the library card catalog. Starred articles are available in the library.

"The purposes of the program being conducted under the present contract are: 1) To determine the feasibility of cladding zirconium on uranium by a direct casting process. 2) To investigate the diffusion of liquid metals in the fusion zone."

Late in 1958 it became apparent that some fuel elements were failing in the Hanford reactors as a result of water entering through the weld. The mode of entry appeared to be first through a void in the weld, then through a non-wet area or a train of voids in the braze, and finally to the uranium core. Defective closures of a similar nature were also typical of many fuel elements which have failed in the autoclaving operation as shown in Figure 1.